TY - JOUR
T1 - Evaluating direct vessel injection accident-event progression of AP1000 and key figures of merit to support the design and development of water-cooled small modular reactors
AU - Abdellatif, Hossam H.
AU - Bhowmik, Palash K.
AU - Arcilesi, David
AU - Sabharwall, Piyush
N1 - Funding Information:
This research was funded by United State (U.S.) Department of Energy (DOE) Advanced Reactor Demonstration Project (ARDP) program office grant number ARDP-20-23819. Funding Opportunity Number DE-FOA-0002271, Risk Reduction Pathway. The authors would like to thank the U.S. DOE National Reactor Innovation Center (NRIC), ARDP program office, and Irradiation Experiment and Thermal Hydraulics Analysis Department at Idaho National Laboratory (INL) for the encouragement and support.
Publisher Copyright:
© 2024 Korean Nuclear Society
PY - 2024
Y1 - 2024
N2 - The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.
AB - The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.
KW - AP1000
KW - Accident management
KW - Accident-event progression
KW - Loss-of-coolant accident
KW - Passive safety systems
KW - Small modular reactor
KW - Small-break loss-of-coolant accident
UR - http://www.scopus.com/inward/record.url?scp=85185776684&partnerID=8YFLogxK
UR - https://www.mendeley.com/catalogue/34df86b7-dc56-3343-a9e7-baf0db15502d/
U2 - 10.1016/j.net.2024.01.049
DO - 10.1016/j.net.2024.01.049
M3 - Article
AN - SCOPUS:85185776684
SN - 1738-5733
JO - Nuclear Engineering and Technology
JF - Nuclear Engineering and Technology
ER -